Neutronic effect of graphite dimensional change in a small modular molten salt reactor

Zhu, Guifeng; Kang, Xuzhong; Zou, Chunyan; Zou, Yang; Yan, Rui; Dai, Ye

doi:10.1002/er.5964

Cited by 13 publications

(8 citation statements)

References 27 publications

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“…As studied in Section 3.2, the neutron loss ratio plays a very critical role in the utilization of fuel under different energy spectrums, and the higher the neutron loss ratio, the lower the utilization of fuel. Neutron losses are mainly due to absorption by structural materials and leakage, and are related to the structural design of the reactor and the size of the core [33,34]. In order to investigate the neutron loss ratio of the reactor at different energy spectra with the core size, the bare reactor model is chosen as the object of study, i.e., a model containing only the active zone of the core without the reflector, BC 4 absorber layer, and metallic structural materials.…”

Section: Heavy Metal Mole Fraction and Molten Salt Volumementioning

confidence: 99%

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Zhu

Liu

et al. 2023

International Journal of Energy Research

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Once-through molten salt reactors have the advantages of easy availability of fuel, nuclear nonproliferation, and low technical difficulty. It is expected to be the earliest commercial molten salt reactor fuel cycle mode. The current research on the neutronic performance of the once-through fuel cycle mainly focuses on the thermal spectrum range, and broader spectrum research needs to be carried out further. In particular, the fuel utilization characteristics of molten salt fast reactors in once-through fuel cycle are not yet clear. In addition, small modular design is one of the trends of molten salt reactor development, and small modular design under thermal and fast spectrum has different advantages and needs further comparative analysis. In this paper, the single lattice, bare reactor, and core with reflector models are used to compare the neutronic performance in once-through fuel cycle from thermal to fast spectrum, which includes burnup, variation of fuel salt composition and volume with burnup, temperature reactivity coefficient, neutron irradiation lifetime, and small modular size. The results show that the burnup increases first, then decreases, and then increases again with the increase of the average lethargy causing fission (EALF). For small-scale molten salt reactors, the difference in fuel utilization under thermal and fast spectra is small. However, for large-scale reactors, the burnup under fast spectrum is significantly higher than that under thermal spectrum. When the EALF is large enough, and the neutron loss ratio is small enough, the breed-and-burn (BNB) fuel cycle mode can be realized, and then, the fuel utilization will be significantly improved. In the fast spectrum reactors, the HN and volume of molten salt change violently because of the longer burnup period. Based on the infinite single lattice model, the temperature reactivity coefficient increases first, then decreases, and increases again with EALF. It is negative over the entire energy spectrum. In addition, the small modular designs are discussed by the bare reactor model, and the neutron irradiation lifetime of the materials is analyzed by the core model with a reflector.

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Section: Heavy Metal Mole Fraction and Molten Salt Volumementioning

confidence: 99%

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Zhu

Liu

et al. 2023

International Journal of Energy Research

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“…In scheme E, we analyzed whether there was a linear relationship between the shielding effects and boron content. In the case of scheme E, the 10 B concentration in Hastelloy varied from 3 wt% to 30 wt%, and the thickness of the side reflector (without boron added) remains unchanged at 35 cm.…”

Section: Modelmentioning

confidence: 99%

“…One of the challenges for SM-MSRs is that the materials such as graphite and Ni-based alloys exposed to high neutron flux have short lifespans and those materials immersed in the highly radioactive fuel salt are not easily replaced. Some studies [9,10] on the evaluation and extension analysis of the lifetime of graphite have been carried out. The Ni-based alloy has more than one order of magnitude lower bearable DPA (displacements per atom) than graphite.…”

Section: Introductionmentioning

confidence: 99%

Neutron/Gamma Radial Shielding Design of Main Vessel in a Small Modular Molten Salt Reactor

Zhu

Zou

et al. 2023

JNE

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The SM-MSR (small modular molten salt reactor) has a good prospect for development with regards to combining the superiority of the molten salt reactor and modularization technologies, showing the advantages of safety, reliability, low economic cost and flexibility of site selection. However, because its internal structural parts are not easily replaced, and the outer shielding structure is limited, the lifespan of the reactor vessel and its in-reactor shielding design needs to be addressed. In order to find an optimal shielding model with both high fuel efficiency and strong radiation shielding capability, five different design schemes were proposed in this work, which varied in thickness and boron concentration in inner-shielding materials. The neutron/gamma flux and DPA (displacements per atom)/helium production rates were evaluated to obtain an appropriate scheme. Several beneficial results were obtained. Considering the above factors and the actual manufacturing process, 20 cm-thick boron graphite with a 5 wt% Boron-10 concentration combined with a 1 cm-thick Hastelloy barrel was chosen as the in-reactor shielding structure. Outside the reactor, the neutron flux was reduced to 8.33 × 1010 cm−2 s−1, and the gamma flux was decreased to 1.13 × 1011 cm−2 s−1. The vessel/barrel material could maintain a lifespan of more than 10 years, while the burnup depth was 6.25% lower than that of a model without inner-shielding. The conclusions of this research can provide important references for the shielding design and parameter selections of small molten salt reactors in the future.

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“…Compared to a solid-fueled reactor, an MSR permits a much broader range of PD in its design benefiting from the liquid fuel form. Even though the increase of PD could cause greater damage to the MSR graphite material due to the increasing neutron flux, 14,15 this problem can be overcome by the replacement of the modular reactor core with its fuel assemblies and by the special design of the fuel assemblies to extend the lifespan of an MSR. Thorcon, 16 which is designed as a small modular MSR, allows a periodic replacement of its critical components including the reactor vessel and fuel salt by gantry cranes.…”

Section: Introductionmentioning

confidence: 99%

Impacts of power density on the breeding performance of molten salt reactors

Chen

Zhang

Zou

et al. 2022

Intl J of Energy Research

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Summary A liquid‐fueled molten salt reactor (MSR) can be flexibly designed with a broad range of power density (PD) and neutron spectrum, both of which have a strong impact on the breeding performance of the fuel cycle. In this work, a thermal single fluid double‐zone Th‐based MSR (SD‐TMSR) and an improved molten salt fast reactor (IMSFR) with different PDs ranging from 40 to 300 MWth/m3 are introduced to evaluate the breeding performance. The results indicate that the breeding ratios for the pure Th‐U (BRTU) and the fissile fuel (BRfis) at equilibrium state are negatively correlated with PD. As the PD increases from 40 to 300 MWth/m3, the BRTU declines by about 9.2% and 5.6% in the SD‐TMSR and IMSFR, respectively, due to the decreases in the probability of the natural decay and extraction of 233Pa in its total vanishment and of the ratio of the 232Th capture rate to the 233U absorption rate. Meanwhile, the BRfis keeps about 6% lower than the BRTU in both the cores because of the larger absorption cross sections of 235U and 239Pu than the capture cross sections of their parent nuclides. Owing to the low capture‐to‐fission ratios of minor actinides (MAs) and the reduced absorption cross sections of fission products, the BRTU and BRfis in the IMSFR keep about 6% higher than those in the SD‐TMSR when PD varies from 40 to 300 MWth/m3. In addition, the net 233U production can achieve the highest value of 60.5 kg/year at 90 MWth/m3 in the SD‐TMSR, and it is 154.9 kg/year at 200 MWth/m3 in the IMSFR. Although the net 233U production in the IMSFR is higher than that in the SD‐TMSR, the doubling time (DT) keeps larger in the IMSFR with the increase of PD from 40 to 100 MWth/m3 due to its higher initial loading of fissile material. In addition, the highest net 233U production of 154.9 kg/year in the IMSFR with a PD of 200 MWth/m3 can be used as the igniting fissile fuel for the SD‐TMSR, in which the corresponding DT is as short as 8.3 years. In conclusion, to achieve a thorium breeding in MSRs with different neutron spectra, a small PD less than 150 MWth/m3 is preferred to achieve a positive breeding gain and a positive net 233U production in the SD‐TMSR, while a higher PD ranging from 150 to 200 MWth/m3 is more attractive to acquire a higher net 233U production of more than 146 kg/year and a shorter DT of less than 35 years in the IMSFR.

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Neutronic effect of graphite dimensional change in a small modular molten salt reactor

Cited by 13 publications

References 27 publications

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Neutron/Gamma Radial Shielding Design of Main Vessel in a Small Modular Molten Salt Reactor

Impacts of power density on the breeding performance of molten salt reactors

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